Development of nuclear data processing code FRENDY version 2

• Published in: Journal of nuclear science and technology Vol. 61; no. 6; pp. 830 - 839
• Main Authors: Tada, Kenichi, Yamamoto, Akio, Kunieda, Satoshi, Konno, Chikara, Kondo, Ryoichi, Endo, Tomohiro, Chiba, Go, Ono, Michitaka, Tojo, Masayuki
• Format: Journal Article
• Language: English
• Published: Tokyo Taylor & Francis Ltd 02.06.2024
• Subjects: Cross-sections; Data processing; Nuclides; Radiation transport; Resonance; Statistical analysis
• ISSN: 0022-3131; 1881-1248
• DOI: 10.1080/00223131.2023.2278600

Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE-formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g. neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6-formatted file. FRENDY version 2 was released including these new functions. It generates GENDF- and MATXS-formatted neutron multi-group cross section files from an ACE-formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code.